<p> The alpha-activity of natural samarium has been studied using an ionization chamber. A gridless ionization chamber has been developed in order to obtain high resolution and high sensitivity. The half lives and the energies of the alpha decay of Sm147 and Sm148 were measured where as the alpha-activity of Sm149 and Sm146 could not be detected.</p> <p> Experiments were carried out in order to search for the existence of Particle stable neutron clusters in the range of Mass 6-10 as a component of a nuclear reactor flux and as the product of high energy proton spallation of heavy nuclei.</p> / Thesis / Doctor of Philosophy (PhD)

A motivação deste trabalho vem da necessidade de agilizar, de maneira segura, o processo dos cálculos neutrônicos e termo-hidráulicos durante a realização da mudança de configuração do reator nuclear de pesquisas IEA-R1. A metodologia de cálculo existente envolve a execução de vários programas computacionais e códigos nucleares nas áreas de física de reatores e termo-hidráulica. Para tanto, foi elaborado uma plataforma de gerenciamento que conjuga os cálculos dos programas já consagrados (\"Two Db\", \"Leopard\", \"Citation\", \"Dens\" e \"Cobra\") numa única ferramenta computacional. A principal contribuição deste trabalho é a concepção de uma interface que simplificará a rotina de cálculos executados no reator IEA-R1, tendo um aspecto didático e uma aparência mais amigável e moderna. / The motivation for this work comes from the need for the process of neutronic and termo-hidraulic calculation to becomes faster, in a safe way, during the configuration change of the research nuclear reactor IEA-R1. The current calculation methodology requires the execution of several softwares and nuclear codes in the areas of Reactor Physics and Thermo-Hidraulic. Therefore, a software was elaborated, which will combines the calculation mentioned in a single computational tool. The main contribution of this work is the creation of an interface that makes the IEA-R1 research reactor executed calculation routine simpler, having didactic aspect e a more friendly and modern appearance.

A motivação deste trabalho vem da necessidade de agilizar, de maneira segura, o processo dos cálculos neutrônicos e termo-hidráulicos durante a realização da mudança de configuração do reator nuclear de pesquisas IEA-R1. A metodologia de cálculo existente envolve a execução de vários programas computacionais e códigos nucleares nas áreas de física de reatores e termo-hidráulica. Para tanto, foi elaborado uma plataforma de gerenciamento que conjuga os cálculos dos programas já consagrados (\"Two Db\", \"Leopard\", \"Citation\", \"Dens\" e \"Cobra\") numa única ferramenta computacional. A principal contribuição deste trabalho é a concepção de uma interface que simplificará a rotina de cálculos executados no reator IEA-R1, tendo um aspecto didático e uma aparência mais amigável e moderna. / The motivation for this work comes from the need for the process of neutronic and termo-hidraulic calculation to becomes faster, in a safe way, during the configuration change of the research nuclear reactor IEA-R1. The current calculation methodology requires the execution of several softwares and nuclear codes in the areas of Reactor Physics and Thermo-Hidraulic. Therefore, a software was elaborated, which will combines the calculation mentioned in a single computational tool. The main contribution of this work is the creation of an interface that makes the IEA-R1 research reactor executed calculation routine simpler, having didactic aspect e a more friendly and modern appearance.

O trabalho teve como objetivo a investigação de Metodologias de Cálculo dos Reatores Subcríticos acionados por fonte externa de nêutrons, tais como, \"Accelerator Driven Subcritical Reactor\" (ADSR) e \"Fusion Driven Subcritical Reator\" (FDSR) , que são reatores nucleares subcríticos com uma fonte externa de nêutrons. Tais nêutrons são produzidos, no caso do ADSR, através da interação de partículas aceleradas (prótons, deutério) com um alvo (Pb, Bi, etc) ou através das reações de fusão, no caso do FDSR. Este conceito de reator vem sendo objeto de intensa pesquisa, sobretudo pela possibilidade de ser utilizado para transmutar o enorme inventario de rejeitos nucleares, principalmente os transurânicos (TRU) e os produtos de fissão de meia-vida longa (LLFP). Neste trabalho enfatiza os seguintes aspectos: (i) complementar e aprimorar a metodologia de cálculos neutrônicos com queima e transmutação e implementá-la computacionalmente; (ii) e utilizando esta metodologia, participar dos Projetos Coordenados de Pesquisa (CRP) da Agência Internacional de energia Atômica \"Analytical and Experimental Benchmark Analysis of ADS\" e \"Collaborative work on use of LEU in ADS\", principalmente na reprodução dos resultados experimentais da instalação subcrítica Yalina Booster e também no cálculo de um núcleo subcrítico do reator IPEN/MB-01, (iii) analisar comparativamente diferentes bibliotecas de dados nucleares, no cálculo de parâmetros integrais (keff), diferenciais (espectro, fluxo) e de queima e transmutação (inventário ao final do ciclo) e (iv) aplicar a metodologia desenvolvida em um estudo que possa ajudar na escolha futura de um sistema transmutador dedicado. Foram utilizados para tanto os seguintes códigos: MCNP (Transporte de partículas por Monte Carlo), MCB (acoplamento do MCNP com código de transmutação) e o sistema NJOY para o processamento dos arquivos de dados nucleares avaliados. / This works had as goal to investigate calculational methodologies on subcritical source driven reactor, such as Accelerator Driven Subcritical Reactor (ADSR) and Fusion Driven Subcritical Reactor (FDSR). Intense R&D has been done about these subcritical concepts, mainly due to Minor Actinides(MA) and Long Lived Fission Products(LLFP) transmutation possibilities. In this work, particular emphasis has been given to: (i ) complement and improve calculation methodology with neutronic transmutation and decay capabilities and implement it computationally, (ii ) utilization of this methodology in the Coordinated Research Project (CRP) of the International Atomic Energy Agency Analytical and Experimental Benchmark Analysis of ADS and in the Collaborative Work on Use of Low Enriched Uranium in ADS, especially in the reproduction of the experimental results of the Yalina Booster subcritical assembly and study of a subcritical core of IPEN/MB-01 reactor, (iii ) to compare dierent nuclear data libraries calculation of integral parameters,such as keff and ksrc, and dierential distributions, such as spectrum and ux, and nuclides inventories and (iv ) apply the developed methodology in a study that may help future choices about dedicated transmutation system. The following tools have been used in this work: MCNP (Monte Carlo N particle transport code), MCB (enhanced version of MCNP that allows burnup calculation) and NJOY to process nuclear data from evaluated nuclear data files.

Neste trabalho, propõe-se um estudo para verificação do impacto da redução na concentração de urânio nas placas laterais dos elementos combustíveis do reator IEA-R1, nas análises neutrônica e termo-hidráulica. Ao se desenvolver o referido trabalho, reproduziu-se estudo conduzido anteriormente pelo IPEN-CNEN/SP, simulando a queima de elementos combustíveis, cujas placas laterais apresentam densidade de urânio reduzida para 50, 60 e 70% em relação às demais placas do elemento combustível. Tal estudo inicia-se com a análise neutrônica, cujo primeiro passo é o cálculo das seções de choque dos materiais presentes no núcleo a partir de suas concentrações iniciais, com a utilização do código computacional HAMMER; o segundo passo é o cálculo dos fluxos de nêutrons dos grupos rápido e térmico e das densidades de potência nos elementos combustíveis estudados em modelagem do núcleo feita no código computacional CITATION, que utiliza os dados gerados pelo HAMMER. Terminada a análise neutrônica e definidos os elementos combustíveis mais críticos com maior densidade de potência, executa-se a análise termo-hidráulica, que utiliza o modelo termo-hidráulico MCTR-IEA-R1, o qual é baseado no pacote comercial EES. A densidade de potência gerada pelo CITATION é utilizada como dado de entrada da análise termo-hidráulica nas equações de balanço de energia do modelo para o cálculo das temperaturas nos pontos de interesse. Neste trabalho, é feita a comparação da operação do reator com três diferentes densidades de urânio nas placas laterais. Concluiu-se que a redução da densidade de urânio contribui para que a temperatura da superfície do revestimento não ultrapasse o limite estabelecido como condição de operação do reator; não há impacto significativo na queima final dos elementos combustíveis, nem na reatividade do reator IEA-R1. A redução de urânio nas placas laterais dos elementos combustíveis do reator IEA-R1 mostrou ser uma opção viável para evitar problemas de corrosão devido a altas temperaturas. / This master thesis presents a study to verify the impact of the uranium concentration reduction in the side plates of the reactor IEA-R1 fuel elements on the neutronic and thermal-hydraulic analyses. To develop such study, a previous IPEN-CNEN/SP research was reproduced by simulating the fuel elements burn-up, with side plate uranium density reduced to 50, 60 and 70% of the standard fuel element plates. This research begins with the neutronic analysis using the computer code HAMMER and the first step consists in the calculation of the cross section of all materials presented at the reactor core, with their initial concentration; the second step consists in the calculation of the fast and thermal neutron group fluxes and power densities for fuel elements using the computer code CITATION. HAMMER output data is used as input data. Once the neutronic analysis is finished and the most critical fuel elements with highest power density have been defined, the thermal-hydraulics analysis begins. This analysis uses MCTR-IEA-R1 thermal-hydraulics model, which equations are solved by commercial code EES. Thermal-hydraulics analysis input is the power density data calculated by CITATION: it is considered the highest power density on each fuel element, where there is a higher energy release and, consequently, higher temperatures. This data is used on energy balance equations to calculate temperatures on critical fuel element regions. Reactor operation comparison for three different uranium densities on fuel side plates is presented. Uranium density reduction contributes to the cladding surface temperature to remain below the established limit, as reactor operation safety requirement and it does not affect significantly fuel element final burn-up nor reactor reactivity. The reduction of uranium in the side plates of the fuel elements of the IEA-R1 showed to be a viable option to avoid corrosion problems due to high temperatures.

La méthode des caractéristiques (méthode MOC) est une méthode efficace et flexible de résolution de l’équation de transport. Cette approche a été considérablement utilisée dans les calculs en deux dimensions car elle permet de traiter des géométries complexes et elle possède un bon ratio temps/précision. Cependant, malgré les améliorations des moyens de stockage et de calcul dans le secteur informatique, un calcul direct en trois dimensions reste encore impossible.Dans ce travail, nous introduisons et analysons plusieurs modifications de la méthode MOC dans le but de réduire la quantité requise de mémoire ainsi que la charge de calcul. Ce document se penche sur l’étude d’une approximation spatiale aux ordres supérieurs pour le flux volumique. En se démarquant de la méthode classique (la méthode MOC constante par morceaux) l’augmentation des détails de la représentation du flux volumique peut permettre de réduire la taille des mailles tout en gardant une bonne précision. Les résultats numériques effectués sur des benchmarks confirment les gains en ratio temps/précision. En ce qui concerne le stockage mémoire, le nombre de trajectoires influe sur la quantité de données à stocker. De ce fait, nous explorons une méthode de traçage par traceurs locaux définis par sous domaines possédant la même géométrie. Les redondances présentes dans les coeurs des réacteurs nucléaires promettent une réduction importante de la quantité de mémoire requise. Deux méthodes de traçage ont été étudiées : la première est une méthode de traçage non-uniforme prenant en compte les discontinuités dans la maille et la deuxième est une méthode fondée sur des trajectoires périodiques et continues d’une maille à l’autre. / The method of characteristics is a flexible and efficient method solving the transport equation. It has been largely used in two dimension calculations because it enables to study complex geometries and it has a good time/precision ratio. However, despite agreat improvement in storage capacities and computing power, a direct three dimension calculation is still unreachable.In the following work, we introduce and analyze several modifications of the methodof characteristics (MOC) in order to reduce the memory usage as well as calculation burden. This document aims at studying a higher order spatial approximation for theflux. It steps away from the classical method (constant MOC) by introducing an increaseof details of the representation of the flux, which may enable to reduce the size of thegrid while keeping a good precision. Numerical results tested on benchmarks show animprovement of time/precision ratio.Regarding the memory storage, the number of trajectories has an influence on the amount of data to be stored. Hence, we study a tracking method based on local tracks defined for all subdomains having the same geometry. Redundancies happening in a reactor core suggest an important reduction of required memory. Two tracking methods have been studied, the first one being a non-uniform tracking method including subdomain discontinuities and the other being a method based on periodic and continuous trajectories for a subdomain to another.

O trabalho teve como objetivo a investigação de Metodologias de Cálculo dos Reatores Subcríticos acionados por fonte externa de nêutrons, tais como, \"Accelerator Driven Subcritical Reactor\" (ADSR) e \"Fusion Driven Subcritical Reator\" (FDSR) , que são reatores nucleares subcríticos com uma fonte externa de nêutrons. Tais nêutrons são produzidos, no caso do ADSR, através da interação de partículas aceleradas (prótons, deutério) com um alvo (Pb, Bi, etc) ou através das reações de fusão, no caso do FDSR. Este conceito de reator vem sendo objeto de intensa pesquisa, sobretudo pela possibilidade de ser utilizado para transmutar o enorme inventario de rejeitos nucleares, principalmente os transurânicos (TRU) e os produtos de fissão de meia-vida longa (LLFP). Neste trabalho enfatiza os seguintes aspectos: (i) complementar e aprimorar a metodologia de cálculos neutrônicos com queima e transmutação e implementá-la computacionalmente; (ii) e utilizando esta metodologia, participar dos Projetos Coordenados de Pesquisa (CRP) da Agência Internacional de energia Atômica \"Analytical and Experimental Benchmark Analysis of ADS\" e \"Collaborative work on use of LEU in ADS\", principalmente na reprodução dos resultados experimentais da instalação subcrítica Yalina Booster e também no cálculo de um núcleo subcrítico do reator IPEN/MB-01, (iii) analisar comparativamente diferentes bibliotecas de dados nucleares, no cálculo de parâmetros integrais (keff), diferenciais (espectro, fluxo) e de queima e transmutação (inventário ao final do ciclo) e (iv) aplicar a metodologia desenvolvida em um estudo que possa ajudar na escolha futura de um sistema transmutador dedicado. Foram utilizados para tanto os seguintes códigos: MCNP (Transporte de partículas por Monte Carlo), MCB (acoplamento do MCNP com código de transmutação) e o sistema NJOY para o processamento dos arquivos de dados nucleares avaliados. / This works had as goal to investigate calculational methodologies on subcritical source driven reactor, such as Accelerator Driven Subcritical Reactor (ADSR) and Fusion Driven Subcritical Reactor (FDSR). Intense R&D has been done about these subcritical concepts, mainly due to Minor Actinides(MA) and Long Lived Fission Products(LLFP) transmutation possibilities. In this work, particular emphasis has been given to: (i ) complement and improve calculation methodology with neutronic transmutation and decay capabilities and implement it computationally, (ii ) utilization of this methodology in the Coordinated Research Project (CRP) of the International Atomic Energy Agency Analytical and Experimental Benchmark Analysis of ADS and in the Collaborative Work on Use of Low Enriched Uranium in ADS, especially in the reproduction of the experimental results of the Yalina Booster subcritical assembly and study of a subcritical core of IPEN/MB-01 reactor, (iii ) to compare dierent nuclear data libraries calculation of integral parameters,such as keff and ksrc, and dierential distributions, such as spectrum and ux, and nuclides inventories and (iv ) apply the developed methodology in a study that may help future choices about dedicated transmutation system. The following tools have been used in this work: MCNP (Monte Carlo N particle transport code), MCB (enhanced version of MCNP that allows burnup calculation) and NJOY to process nuclear data from evaluated nuclear data files.

Neste trabalho, propõe-se um estudo para verificação do impacto da redução na concentração de urânio nas placas laterais dos elementos combustíveis do reator IEA-R1, nas análises neutrônica e termo-hidráulica. Ao se desenvolver o referido trabalho, reproduziu-se estudo conduzido anteriormente pelo IPEN-CNEN/SP, simulando a queima de elementos combustíveis, cujas placas laterais apresentam densidade de urânio reduzida para 50, 60 e 70% em relação às demais placas do elemento combustível. Tal estudo inicia-se com a análise neutrônica, cujo primeiro passo é o cálculo das seções de choque dos materiais presentes no núcleo a partir de suas concentrações iniciais, com a utilização do código computacional HAMMER; o segundo passo é o cálculo dos fluxos de nêutrons dos grupos rápido e térmico e das densidades de potência nos elementos combustíveis estudados em modelagem do núcleo feita no código computacional CITATION, que utiliza os dados gerados pelo HAMMER. Terminada a análise neutrônica e definidos os elementos combustíveis mais críticos com maior densidade de potência, executa-se a análise termo-hidráulica, que utiliza o modelo termo-hidráulico MCTR-IEA-R1, o qual é baseado no pacote comercial EES. A densidade de potência gerada pelo CITATION é utilizada como dado de entrada da análise termo-hidráulica nas equações de balanço de energia do modelo para o cálculo das temperaturas nos pontos de interesse. Neste trabalho, é feita a comparação da operação do reator com três diferentes densidades de urânio nas placas laterais. Concluiu-se que a redução da densidade de urânio contribui para que a temperatura da superfície do revestimento não ultrapasse o limite estabelecido como condição de operação do reator; não há impacto significativo na queima final dos elementos combustíveis, nem na reatividade do reator IEA-R1. A redução de urânio nas placas laterais dos elementos combustíveis do reator IEA-R1 mostrou ser uma opção viável para evitar problemas de corrosão devido a altas temperaturas. / This master thesis presents a study to verify the impact of the uranium concentration reduction in the side plates of the reactor IEA-R1 fuel elements on the neutronic and thermal-hydraulic analyses. To develop such study, a previous IPEN-CNEN/SP research was reproduced by simulating the fuel elements burn-up, with side plate uranium density reduced to 50, 60 and 70% of the standard fuel element plates. This research begins with the neutronic analysis using the computer code HAMMER and the first step consists in the calculation of the cross section of all materials presented at the reactor core, with their initial concentration; the second step consists in the calculation of the fast and thermal neutron group fluxes and power densities for fuel elements using the computer code CITATION. HAMMER output data is used as input data. Once the neutronic analysis is finished and the most critical fuel elements with highest power density have been defined, the thermal-hydraulics analysis begins. This analysis uses MCTR-IEA-R1 thermal-hydraulics model, which equations are solved by commercial code EES. Thermal-hydraulics analysis input is the power density data calculated by CITATION: it is considered the highest power density on each fuel element, where there is a higher energy release and, consequently, higher temperatures. This data is used on energy balance equations to calculate temperatures on critical fuel element regions. Reactor operation comparison for three different uranium densities on fuel side plates is presented. Uranium density reduction contributes to the cladding surface temperature to remain below the established limit, as reactor operation safety requirement and it does not affect significantly fuel element final burn-up nor reactor reactivity. The reduction of uranium in the side plates of the fuel elements of the IEA-R1 showed to be a viable option to avoid corrosion problems due to high temperatures.

<p dir="ltr">In the last few decades, the US Department of Energy established the Generation-IV Initiative to advance the design of nuclear energy systems with a focus on fast nuclear reactors. Special interest has been placed on Sodium-cooled Fast Reactors (SFRs) along with metallic fuels. SFRs are attractive because of their ability to utilize fast neutrons effectively, allowing for efficient transmutation of long-lived radioactive isotopes and a more complete use of fissile material. This capability significantly reduces nuclear waste and improves fuel sustainability compared to other Generation IV reactors. The metallic fuels, such as U-Zr and U-Pu-Zr, are attractive because of their higher thermal conductivity and higher density of fissile material, leading to improved breeding ratios and higher burnup rates. Through years of testing, SFRs, such as the EBR-II, could achieve a very high burnup (up to 750 GWD/tU) while traditional generation I to III+ reactors achieve around 30 GWD/tU. Since fast reactors, particularly SFRs, operate on a hard neutron spectrum, they utilize different geometries and cooling materials. This requires the use of different mechanistic models in nuclear codes to accurately capture the underlying physics. The NRC has recently shown interest in upgrading its diffusion codes to support the integration of SFRs. They have shown additional interest in improving the simulation capability of SFR metallic fuel, specifically U-10wt%Zr, even for high fuel burnups in excess of 10 at.%. </p><p dir="ltr">The purpose of this thesis is multifaceted. It serves to develop new mechanistic modelsto support the modeling and analysis of SFRs in existing nodal codes; it also serves to advance the current understanding of the principal effects of U-Zr fuel behavior during steady-state conditions. From the neutronics perspective, a new nodal method will be developed within the PARCS nodal code, along with generalized concepts such as reactivity feedback coefficients and thermal expansion, which will then be validated against the EBRII steady-state benchmark. From the fuel behavior perspective, mechanistic models that describe fuel redistribution, temperature distribution, fission gas, mechanical stresses, and point defect generation will be developed into the Purdue Fuel Performance (PFP) code for U-Zr fuel. The code will then be validated against the radial fuel concentration profile of two EBR-II U-Zr fuel pins.</p>

La thèse a essentiellement été motivée par la volonté croissante de maîtriser les incertitudes des données nucléaires, pour des raisons de sûreté nucléaire. Elle vise en particulier les sections efficaces indispensables aux calculs neutroniques des réacteurs rapides au sodium de Génération IV (RNR-Na), et les moyens permettant de les évaluer.Le principal objectif de la thèse est de fournir et montrer l’intérêt de nouveaux outils permettant de réaliser des évaluations cohérentes, avec des incertitudes maîtrisées et fiables. Pour répondre aux attentes, différentes méthodes ont été implémentées dans le cadre du code CONRAD, développé au CEA de Cadarache, au Département d’Étude des Réacteurs.Après l’état des lieux et la présentation des différents éléments nécessaires pour effectuer une évaluation, il est présenté des résolutions stochastiques de l’inférence Bayésienne. Elles permettent de fournir d’une part, des informations supplémentaires à l’évaluateur par rapport à la résolution analytique et d’autre part, de valider cette dernière. Les algorithmes ont été testés avec succès à travers plusieurs cas, malgré des temps de calcul plus longs faute aux méthodes de type Monte Carlo.Ensuite, ce travail a rendu possible, dans CONRAD, de prendre en compte des contraintes dites microscopiques. Elles sont définies par l’ajout ou le traitement d’informations additionnelles par rapport à l’évaluation traditionnelle. Il a été développé un algorithme basé sur le formalisme des multiplicateurs de Lagrange pour résoudre les problèmes de continuité entre deux domaines en énergies traitées par deux théories différentes. De plus, d’autres approches sont présentées, avec notamment l’utilisation de la marginalisation, permettant soit de compléter une évaluation existante en ajoutant des matrices de covariance, soit de considérer une incertitude systématique pour une expérience décrite par deux théories. Le bon fonctionnement des différentes méthodes implémentées est illustré par des exemples, dont celui de la section efficace totale de l’238U.Enfin, les dernières parties de la thèse se focalisent sur le retour des expériences intégrales, par méthodes d’assimilation de données intégrales. Cela permet de réduire les incertitudes sur les sections efficaces d’intérêt pour les réacteurs rapides. Ce document se clôt par la présentation de quelques résultats clefs sur les sections efficaces de l’238U et du 239Pu, avec la considération d’expériences comme PROFIL et PROFIL-2 dans Phénix ou encore Jezebel. / The thesis has been motivated by a wish to increase the uncertainty knowledge on nuclear data, for safety criteria. It aims the cross sections required by core calculation for sodium fast reactors (SFR), and new tools to evaluate its.The main objective of this work is to provide new tools in order to create coherent evaluated files, with reliable and mastered uncertainties. To answer those problematic, several methods have been implemented within the CONRAD code, which is developed at CEA of Cadarache.After a summary of all the elements required to understand the evaluation world, stochastic methods are presented in order to solve the Bayesian inference. They give the evaluator more information about probability density and they also can be used as validation tools. The algorithms have been successfully tested, despite long calculation time.Then, microscopic constraints have been implemented in CONRAD. They are defined as new information that should be taken into account during the evaluation process. An algorithm has been developed in order to solve, for example, continuity issues between two energy domains, with the Lagrange multiplier formalism. Another method is given by using a marginalization procedure, in order to either complete an existing evaluation with new covariance or add systematic uncertainty on an experiment described by two theories. The algorithms are well performed along examples, such the 238U total cross section.The last parts focus on the integral data feedback, using methods of integral data assimilation to reduce the uncertainties on cross sections. This work ends with uncertainty reduction on key nuclear reactions, such the capture and fission cross sections of 238U and 239Pu, thanks to PROFIL and PROFIL-2 experiments in Phénix and the Jezebel benchmark.